[Federal Register Volume 83, Number 1 (Tuesday, January 2, 2018)]
[Notices]
[Pages 161-175]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2017-27930]
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NUCLEAR REGULATORY COMMISSION
[NRC-2017-0238]
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses and Combined Licenses Involving No Significant
Hazards Considerations
AGENCY: Nuclear Regulatory Commission.
ACTION: Biweekly notice.
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SUMMARY: Pursuant to Section 189a. (2) of the Atomic Energy Act of
1954, as amended (the Act), the U.S. Nuclear Regulatory Commission
(NRC) is publishing this regular biweekly notice. The Act requires the
Commission to publish notice of any amendments issued, or proposed to
be issued, and grants the Commission the authority to issue and make
immediately effective any amendment to an operating license or combined
license, as applicable, upon a determination by the Commission that
such amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued, from December 5, 2017, to December 18, 2017. The
last biweekly notice was published on December 19, 2017.
DATES: Comments must be filed by February 1, 2018. A request for a
hearing must be filed by March 5, 2018.
ADDRESSES: You may submit comments by any of the following methods
(unless this document describes a different method for submitting
comments on a specific subject):
Federal Rulemaking Website: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0238. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected]. For technical questions, contact
the individual listed in the FOR FURTHER INFORMATION CONTACT section of
this document.
Mail comments to: May Ma, Office of Administration, Mail
Stop: OWFN-2-A13, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001.
For additional direction on obtaining information and submitting
comments, see ``Obtaining Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
FOR FURTHER INFORMATION CONTACT: Lynn Ronewicz, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC
20555-0001; telephone: 301-415-1927, email: [email protected].
SUPPLEMENTARY INFORMATION:
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2017-0238, facility name, unit
number(s), plant docket number, application date, and subject when
contacting the NRC about the availability of information for this
action. You may obtain publicly-available information related to this
action by any of the following methods:
Federal Rulemaking Website: Go to http://www.regulations.gov and search for Docket ID NRC-2017-0238.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly-available documents online in the
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS,
please contact the NRC's Public Document Room (PDR) reference staff at
1-800-397-4209, 301-415-4737, or by email to [email protected]. The
ADAMS accession number for each document referenced (if it is available
in ADAMS) is provided the first time that it is mentioned in this
document.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
B. Submitting Comments
Please include Docket ID NRC-2017-0238, facility name, unit
number(s), plant docket number, application date, and subject in your
comment submission.
The NRC cautions you not to include identifying or contact
information that you do not want to be publicly disclosed in your
comment submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into
ADAMS. The NRC does not routinely edit comment submissions to remove
identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment submissions available
to the public or entering the comment into ADAMS.
II. Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses and Combined Licenses and Proposed No Significant
Hazards Consideration Determination
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in Sec. 50.92 of title 10 of the Code of
Federal Regulations (10 CFR), this means that operation of the facility
in accordance with the proposed amendment would not (1) involve a
significant increase in the probability or consequences of an accident
previously evaluated, or (2) create the possibility of a new or
different kind of accident from any accident previously evaluated; or
(3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be
[[Page 162]]
considered in making any final determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period if circumstances change during the 30-day comment
period such that failure to act in a timely way would result, for
example in derating or shutdown of the facility. If the Commission
takes action prior to the expiration of either the comment period or
the notice period, it will publish in the Federal Register a notice of
issuance. If the Commission makes a final no significant hazards
consideration determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
A. Opportunity To Request a Hearing and Petition for Leave To Intervene
Within 60 days after the date of publication of this notice, any
persons (petitioner) whose interest may be affected by this action may
file a request for a hearing and petition for leave to intervene
(petition) with respect to the action. Petitions shall be filed in
accordance with the Commission's ``Agency Rules of Practice and
Procedure'' in 10 CFR part 2. Interested persons should consult a
current copy of 10 CFR 2.309. The NRC's regulations are accessible
electronically from the NRC Library on the NRC's website at http://www.nrc.gov/reading-rm/doc-collections/cfr/. Alternatively, a copy of
the regulations is available at the NRC's Public Document Room, located
at One White Flint North, Room O1-F21, 11555 Rockville Pike (first
floor), Rockville, Maryland 20852. If a petition is filed, the
Commission or a presiding officer will rule on the petition and, if
appropriate, a notice of a hearing will be issued.
As required by 10 CFR 2.309(d) the petition should specifically
explain the reasons why intervention should be permitted with
particular reference to the following general requirements for
standing: (1) The name, address, and telephone number of the
petitioner; (2) the nature of the petitioner's right under the Act to
be made a party to the proceeding; (3) the nature and extent of the
petitioner's property, financial, or other interest in the proceeding;
and (4) the possible effect of any decision or order which may be
entered in the proceeding on the petitioner's interest.
In accordance with 10 CFR 2.309(f), the petition must also set
forth the specific contentions which the petitioner seeks to have
litigated in the proceeding. Each contention must consist of a specific
statement of the issue of law or fact to be raised or controverted. In
addition, the petitioner must provide a brief explanation of the bases
for the contention and a concise statement of the alleged facts or
expert opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to the specific sources and
documents on which the petitioner intends to rely to support its
position on the issue. The petition must include sufficient information
to show that a genuine dispute exists with the applicant or licensee on
a material issue of law or fact. Contentions must be limited to matters
within the scope of the proceeding. The contention must be one which,
if proven, would entitle the petitioner to relief. A petitioner who
fails to satisfy the requirements at 10 CFR 2.309(f) with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene.
Parties have the opportunity to participate fully in the conduct of the
hearing with respect to resolution of that party's admitted
contentions, including the opportunity to present evidence, consistent
with the NRC's regulations, policies, and procedures.
Petitions must be filed no later than 60 days from the date of
publication of this notice. Petitions and motions for leave to file new
or amended contentions that are filed after the deadline will not be
entertained absent a determination by the presiding officer that the
filing demonstrates good cause by satisfying the three factors in 10
CFR 2.309(c)(1)(i) through (iii). The petition must be filed in
accordance with the filing instructions in the ``Electronic Submissions
(E-Filing)'' section of this document.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to establish when the hearing is held. If the final determination is
that the amendment request involves no significant hazards
consideration, the Commission may issue the amendment and make it
immediately effective, notwithstanding the request for a hearing. Any
hearing would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, then any hearing held would take place before
the issuance of the amendment unless the Commission finds an imminent
danger to the health or safety of the public, in which case it will
issue an appropriate order or rule under 10 CFR part 2.
A State, local governmental body, Federally-recognized Indian
Tribe, or agency thereof, may submit a petition to the Commission to
participate as a party under 10 CFR 2.309(h)(1). The petition should
state the nature and extent of the petitioner's interest in the
proceeding. The petition should be submitted to the Commission no later
than 60 days from the date of publication of this notice. The petition
must be filed in accordance with the filing instructions in the
``Electronic Submissions (E-Filing)'' section of this document, and
should meet the requirements for petitions set forth in this section,
except that under 10 CFR 2.309(h)(2) a State, local governmental body,
or federally recognized Indian Tribe, or agency thereof does not need
to address the standing requirements in 10 CFR 2.309(d) if the facility
is located within its boundaries. Alternatively, a State, local
governmental body, Federally-recognized Indian Tribe, or agency thereof
may participate as a non-party under 10 CFR 2.315(c).
If a hearing is granted, any person who is not a party to the
proceeding and is not affiliated with or represented by a party may, at
the discretion of the presiding officer, be permitted to make a limited
appearance pursuant to the provisions of 10 CFR 2.315(a). A person
making a limited appearance may make an oral or written statement of
his or her position on the issues but may not otherwise participate in
the proceeding. A limited appearance may be made at any session of the
hearing or at any prehearing conference, subject to the limits and
conditions as may be imposed by the presiding officer. Details
regarding the opportunity to make a limited appearance will be provided
by the presiding officer if such sessions are scheduled.
B. Electronic Submissions (E-Filing)
All documents filed in NRC adjudicatory proceedings, including a
request for hearing and petition for
[[Page 163]]
leave to intervene (petition), any motion or other document filed in
the proceeding prior to the submission of a request for hearing or
petition to intervene, and documents filed by interested governmental
entities that request to participate under 10 CFR 2.315(c), must be
filed in accordance with the NRC's E-Filing rule (72 FR 49139; August
28, 2007, as amended at 77 FR 46562, August 3, 2012). The E-Filing
process requires participants to submit and serve all adjudicatory
documents over the internet, or in some cases to mail copies on
electronic storage media. Detailed guidance on making electronic
submissions may be found in the Guidance for Electronic Submissions to
the NRC and on the NRC website at http://www.nrc.gov/site-help/e-submittals.html. Participants may not submit paper copies of their
filings unless they seek an exemption in accordance with the procedures
described below.
To comply with the procedural requirements of E-Filing, at least 10
days prior to the filing deadline, the participant should contact the
Office of the Secretary by email at [email protected], or by
telephone at 301-415-1677, to (1) request a digital identification (ID)
certificate, which allows the participant (or its counsel or
representative) to digitally sign submissions and access the E-Filing
system for any proceeding in which it is participating; and (2) advise
the Secretary that the participant will be submitting a petition or
other adjudicatory document (even in instances in which the
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the
Secretary will establish an electronic docket for the hearing in this
proceeding if the Secretary has not already established an electronic
docket.
Information about applying for a digital ID certificate is
available on the NRC's public website at http://www.nrc.gov/site-help/e-submittals/getting-started.html. Once a participant has obtained a
digital ID certificate and a docket has been created, the participant
can then submit adjudicatory documents. Submissions must be in Portable
Document Format (PDF). Additional guidance on PDF submissions is
available on the NRC's public website at http://www.nrc.gov/site-help/electronic-sub-ref-mat.html. A filing is considered complete at the
time the document is submitted through the NRC's E-Filing system. To be
timely, an electronic filing must be submitted to the E-Filing system
no later than 11:59 p.m. Eastern Time on the due date. Upon receipt of
a transmission, the E-Filing system time-stamps the document and sends
the submitter an email notice confirming receipt of the document. The
E-Filing system also distributes an email notice that provides access
to the document to the NRC's Office of the General Counsel and any
others who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
document on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before adjudicatory documents are
filed so that they can obtain access to the documents via the E-Filing
system.
A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC's Electronic
Filing Help Desk through the ``Contact Us'' link located on the NRC's
public website at http://www.nrc.gov/site-help/e-submittals.html, by
email to [email protected], or by a toll-free call at 1-866-672-
7640. The NRC Electronic Filing Help Desk is available between 9 a.m.
and 6 p.m., Eastern Time, Monday through Friday, excluding government
holidays.
Participants who believe that they have a good cause for not
submitting documents electronically must file an exemption request, in
accordance with 10 CFR 2.302(g), with their initial paper filing
stating why there is good cause for not filing electronically and
requesting authorization to continue to submit documents in paper
format. Such filings must be submitted by: (1) First class mail
addressed to the Office of the Secretary of the Commission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemaking and Adjudications Staff; or (2) courier, express mail, or
expedited delivery service to the Office of the Secretary, 11555
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and
Adjudications Staff. Participants filing adjudicatory documents in this
manner are responsible for serving the document on all other
participants. Filing is considered complete by first-class mail as of
the time of deposit in the mail, or by courier, express mail, or
expedited delivery service upon depositing the document with the
provider of the service. A presiding officer, having granted an
exemption request from using E-Filing, may require a participant or
party to use E-Filing if the presiding officer subsequently determines
that the reason for granting the exemption from use of E-Filing no
longer exists.
Documents submitted in adjudicatory proceedings will appear in the
NRC's electronic hearing docket which is available to the public at
https://adams.nrc.gov/ehd, unless excluded pursuant to an order of the
Commission or the presiding officer. If you do not have an NRC-issued
digital ID certificate as described above, click cancel when the link
requests certificates and you will be automatically directed to the
NRC's electronic hearing dockets where you will be able to access any
publicly available documents in a particular hearing docket.
Participants are requested not to include personal privacy information,
such as social security numbers, home addresses, or personal phone
numbers in their filings, unless an NRC regulation or other law
requires submission of such information. For example, in some
instances, individuals provide home addresses in order to demonstrate
proximity to a facility or site. With respect to copyrighted works,
except for limited excerpts that serve the purpose of the adjudicatory
filings and would constitute a Fair Use application, participants are
requested not to include copyrighted materials in their submission.
For further details with respect to these license amendment
applications, see the application for amendment which is available for
public inspection in ADAMS and at the NRC's PDR. For additional
direction on accessing information related to this document, see the
``Obtaining Information and Submitting Comments'' section of this
document.
Dominion Nuclear Connecticut, Inc. (DNC), Docket No. 50-336, Millstone
Power Station, Unit No. 2, New London County, Connecticut
Date of amendment request: October 4, 2017. A publicly-available
version is in ADAMS under Accession No. ML17284A179.
Description of amendment request: The amendment would revise the
Millstone Power Station, Unit No. 2 (MPS2) Technical Specification (TS)
6.19, ``Containment Leakage Rate Testing Program,'' by replacing the
reference to Regulatory Guide (RG) 1.163, ``Performance-Based
Containment Leak-Test Program,'' with a reference to Nuclear Energy
Institute (NEI) Topical Report NEI 94-01, Revision 3-A, ``Industry
Guideline for Implementing Performance-Based Option of 10 CFR part 50,
Appendix J,'' and the limitations and conditions specified in NEI 94-
01, Revision 2-A, as the
[[Page 164]]
implementing documents used to develop the MPS2 performance-based
leakage testing program in accordance with 10 CFR, Appendix J, Option
B, ``Primary Reactor Containment Leakage Testing for Water-Cooled Power
Reactors.'' The amendment would allow DNC to extend the Type A primary
containment integrated leak rate test interval (ILRT) for MPS2 to 15
years and the Type C local leak rate test interval to 75 months, and
incorporates the regulatory positions stated in RG 1.163.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment involves changes to the MPS2 Containment
Leakage Rate Testing Program. The proposed amendment does not
involve a physical change to the plant or a change in the manner in
which the plant is operated or controlled. The primary containment
function is to provide an essentially leak tight barrier against the
uncontrolled release of radioactivity to the environment for
postulated accidents. As such, the containment and the testing
requirements to periodically demonstrate the integrity of the
containment exist to ensure the plant's ability to mitigate the
consequences of an accident, and do not involve any accident
precursors or initiators.
Therefore, the probability of occurrence of an accident
previously evaluated is not significantly increased by the proposed
amendment.
The proposed amendment adopts the NRC-accepted guidelines of NEI
94-01, Revision 3-A, and the limitations and conditions specified in
NEI 94-01, Rev. 2-A, for development of the MPS2 performance-based
leakage testing program. Implementation of these guidelines
continues to provide adequate assurance that during design basis
accidents, the primary containment and its components will limit
leakage rates to less than the values assumed in the plant safety
analyses. The potential consequences of extending the ILRT interval
to 15 years have been evaluated by analyzing the resulting changes
in risk. The increase in risk in terms of person-rem [roentgen
equivalent man] per year within 50 miles resulting from design basis
accidents was estimated to be acceptably small and determined to be
within the guidelines published in RG 1.174. Additionally, the
proposed change maintains defense-in-depth by preserving a
reasonable balance among prevention of core damage, prevention of
containment failure, and consequence mitigation. DNC has determined
that the increase in Conditional Containment Failure Probability due
to the proposed change is very small.
Therefore, [the proposed change does not involve a significant
increase in the probability or consequences] of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed amendment adopts the NRC-accepted guidelines of NEI
94-01, Revision 3-A, and the limitations and conditions specified in
NEI 94-01, Rev. 2-A, for development of the MPS2 performance-based
leakage testing program, and establishes a 15-year interval for Type
A testing and an interval not to exceed 75 months for Type C
testing. The containment and the testing requirements to
periodically demonstrate the integrity of the containment exist to
ensure the plant's ability to mitigate the consequences of an
accident; do not involve any accident precursors or initiators. The
proposed change does not involve a physical change to the plant
(i.e., no new or different type of equipment will be installed) or a
change to the manner in which the plant is operated or controlled.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in the margin of safety?
Response: No.
The proposed amendment adopts the NRC-accepted guidelines of NEI
94-01, Revision 3-A, and the limitations and conditions specified in
NEI 94-01, Rev. 2-A, for the development of the MPS2 performance-
based leakage testing program, and establishes a 15-year interval
for Type A testing and an interval not to exceed 75 months for Type
C testing. This amendment does not alter the manner in which safety
limits, limiting safety system setpoints, or limiting conditions for
operation are determined. The specific requirements and conditions
of the Containment Leakage Rate Testing Program, as defined in the
TS, ensure that the degree of primary containment structural
integrity and leak-tightness that is considered in the plant's
safety analysis is maintained. The overall containment leakage rate
limit specified by the TS is maintained, and the Type A, Type B, and
Type C containment leakage tests will be performed at the
frequencies established in accordance with the NRC-accepted
guidelines of NEI 94-01, Revision 3-A, and the limitations and
conditions specified in NEI 94-01, Rev. 2-A.
Containment inspections performed in accordance with other plant
programs serve to provide a high degree of assurance that the
containment will not degrade in a manner that is not detectable by
an ILRT. A risk assessment using the current MPS2 PRA [probabilistic
risk assessment] model concluded that extending the ILRT test
interval from 10 years to 15 years results in a small change to the
MPS2 risk profile.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion
Energy, Inc., 120 Tredegar Street, RS-2, Richmond, VA 23219.
NRC Branch Chief: James G. Danna.
DTE Electric Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: August 24, 2017. A publicly-available
version is in ADAMS under Accession No. ML17237A176.
Description of amendment request: The proposed amendment revises
Technical Specification (TS) 3.3.1.1, ``Reactor Protection System (RPS)
Instrumentation,'' to eliminate the main steam line radiation monitor
(MSLRM) functions for initiating (1) a reactor protection system
automatic reactor trip and (2) the associated (Group 1) primary
containment isolation system (PCIS) isolations, which include automatic
closure of the main steam isolation valves (MSIV) and main steam line
(MSL) drain valves. The proposed changes also remove requirements for
Group 1 PCIS isolation from TS 3.3.6.1, ``Primary Containment Isolation
Instrumentation.'' This submittal also proposes the addition of two new
TS Limiting Conditions for Operation, 3.3.7.2 and 3.3.7.3, for the
mechanical vacuum pump and gland seal exhauster trip instrumentation
that will be required to actuate in response to high MSL radiation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes eliminate the MSLRM trip and isolation
functions from initiating an automatic reactor scram and automatic
closure of the MSIVs. The justification for eliminating the MSLRM
trip and MSIV isolation functions is based on the NRC-approved
evaluation provided in GE LTR [General Electric Licensing Topical
Report] NEDO-31400A, ``Safety Evaluation for Eliminating the Boiling
Water Reactor Main Steam Line Isolation Valve Closure Function and
Scram Function of the Main Steam Line Radiation Monitor,'' dated
October 1992.
[[Page 165]]
The MSLRM high radiation RPS scram function has never been
credited to shut down the reactor in response to a postulated CRDA
[control rod drop accident]; instead, the neutron monitoring system
will continue to be the credited means to shut down the reactor in
response to the high flux condition that results from the reactivity
inserted by the CRDA.
The consequences of an accident previously evaluated, have been
re-evaluated consistent with RG [Regulatory Guide] 1.183 Rev. 0 AST
[alternate source term] (10 CFR 50.67) for the applicable DBA
[design basis accident] (i.e., the CRDA) as stipulated in NEDO-
31400A. The supporting dose analyses demonstrate that, with
continued credit for the automatic trip/isolation of the MVPs
[mechanical vacuum pump] as well as a new proposed automatic trip of
the GSEs [gland seal exhauster], the consequences of the accident
are within the regulatory acceptance criteria recommended in RG
1.183 Rev. 0 for compliance with 10 CFR 50.67. As a result, the
consequences of any accident previously evaluated are not
significantly increased.
The proposed modification of the trip logic for the MVPs to
utilize the safety-related MSLRM signals is an improvement over the
current licensed configuration of the MVP trip, which utilizes the
nonsafety-related offgas 2-minute delay pipe radiation monitor
``High-High'' radiation signal. Reliance on the safety-related MSLRM
signal is consistent with similar approved license amendments and,
in addition to improving the quality and reliability of the sensing
circuit, ensures the signal is generated at the time of earliest
possible detection and therefore improves the effectiveness of the
actuation. The trip setpoint utilized corresponds to the same value
previously assigned for initiating MSIV isolation in response to the
design basis CRDA. The offgas 2-minute delay pipe radiation monitor
alarm function is being retained, with a more conservative setpoint,
to continue to provide indication of increased radiation.
Similar to the MVPs, the proposed new trip of the nonsafety-
related GSEs is also necessary to ensure calculated radiological
consequences remain within the regulatory acceptance limits.
Reliance on the safety-related MSLRM signal is consistent with BWR
[boiling water reactor] design for reliable tripping of the
nonsafety-related MVPs and ensures the signal is reliably generated
at the time of earliest possible detection and maximizes the
effectiveness of the actuation.
The proposed changes also include the elimination of the MSLRM
isolation function from automatically closing the MSL drain valves.
The contents of the MSL drain lines are conveyed to the main
condenser. The evaluation of the condenser release path assumes that
100% of CRDA activity released is transported to the main condenser
in 1 second, and therefore, the transportation of the post-CRDA
activity from the reactor coolant to the main condenser either via
MSLs or MSL drain lines is inconsequential and is supported by the
dose analyses performed in support of this submittal.
Neither the MSLRMs nor the MVPs are postulated initiators of any
accident previously evaluated. None of the proposed changes alter
the probability of the occurrence of the CRDA initiating event.
The loss of the GSEs is a malfunction of equipment considered in
UFSAR [updated final safety analysis report] Section 15.12
``Malfunction of Turbine Gland Sealing System.'' In the event that
the operating blower malfunctions, the backup blower will
automatically assume the gas removal requirements. Assuming loss of
both blowers, vacuum will be lost in the gland steam condenser. No
cladding perforations result from a malfunction of the turbine gland
sealing system. The pressure in the gland steam exhaust header will
increase to greater than atmospheric, allowing sealing steam to
escape into the turbine building. If exhauster vacuum falls below a
specified value, caused for example by loss of alternating current
(AC) power, a vacuum switch initiates the closing of the live steam
supply to the gland steam header. Above 50% to 60% reactor power,
the turbine is self-sealing; hence, the packing lines would remain
pressurized under normal operating conditions.
The logic associated with the new trip of the GSEs will be
designed to preserve the existing ability of the backup exhauster to
automatically respond to a loss of the operating exhauster, in the
absence of a valid high MSL radiation trip signal. Similar to the
design of the RPS trip logic that is proposed to be eliminated, the
GSE trip logic will be configured such that no single failure of a
MSLRM can generate a GSE trip signal. As specified in the
``Applicability'' section for the new proposed LCO [limiting
condition for operation] 3.3.7.3, the trip logic will be
automatically bypassed when reactor power is above 10% RTP [rated
thermal power] when the consequences postulated in association with
a CRDA are not credible. On the basis of the configuration of the
GSE trip logic, the quality of the initiating trip logic signal, and
the short duration of normal operation for which the GSE trip logic
will be active, the probability of a malfunction of equipment
leading to the loss of the turbine gland sealing system is not
significantly increased.
The proposed changes do not increase system or component
pressures, temperatures, or flowrates for systems designed to
prevent accidents or mitigate the consequences of an accident. Since
these conditions do not change, the probability of a process-induced
failure or malfunction of a SSC [system, structure, or component] is
not increased.
The addition of MVP and GSE SRs [surveillance requirements] and
LCOs to the TS enhances the reliability of these design functions by
establishing administrative requirements for periodic verification
of their operability.
The reliance on a lower assigned MSL high radiation alarm
setpoint of 1.5 times the full power N-16 background will direct the
control room operators to diagnose and act to mitigate conditions
associated with fuel damage and release sooner than the current
alarm condition which will reduce the potential consequences of a
postulated release due to a CRDA.
On the basis of the above considerations, the proposed changes
do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not increase system or component
pressures, temperatures, or flowrates. Since these conditions do not
change, the likelihood of a process-induced failure or malfunction
of a SSC not previously considered is not increased.
The reliance on the MVP trip to ensure acceptable dose
consequences following a postulated CRDA is consistent with the
original plant design and licensing bases. The re-assignment of the
initiating input for the MVP trip logic to the MSLRM improves the
quality and reliability of the credited trip initiating logic by
relying on safety-related, redundant components. The quality of the
nonsafety-related trip circuit itself is unchanged.
The reliance on the proposed trip of the GSEs is a function that
is credited to ensure acceptable dose consequences following a
postulated CRDA. The use of the safety-related redundant MSLRM
signals and nonsafety-related trip circuit provides the same level
of quality and reliability of the initiating trip logic and trip
circuitry credited to trip the MVPs. These requirements provide the
reliability necessary to ensure the assumptions of the analyzed CRDA
remain valid.
Both the safety-related trip logic and the nonsafety-related
trip circuits associated with the MVP and GSE trips will be designed
to include qualified electrical isolation necessary to ensure the
nonsafety-related trip circuitry cannot induce failures of or affect
the reliability of the safety-related trip logic.
The new GSE trip will be designed to preserve the existing
function for auto-start of the standby exhauster in the event that
the plant experiences a loss of the operating exhauster, in the
absence of a valid high MSL radiation trip signal. An installed
automatic bypass of the GSE trip is actuated once steam flow and
feedwater flow correspond to the same Low Power Setpoint used to
disable the rod block function of the Rod Worth Minimizer during
plant startup. This bypass will minimize the potential for the plant
to experience a loss of both GSEs and potential ensuing turbine trip
due to a failure of the new trip circuit. The status of the GSE trip
bypass will be available to the control room operators and be
required to be verified as a part of the plant general operating
procedures for startup/shutdown.
Adding requirements for the MVP and GSE trip instrumentation in
the TS will ensure that appropriate measures and requirements are in
place such that any release of radioactive material released from a
gross fuel failure will be contained in the main condenser and
processed through the offgas system in the manner credited in the
plant analysis of the CRDA.
The MSLRM trip and isolation functions being eliminated as
described above are only applicable to the CRDA and no other event
[[Page 166]]
in the safety analysis. The proposed changes are consistent with the
revised safety analysis assumptions for a CRDA as described in this
license amendment request.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes eliminating the MSLRM trip and isolation
functions from initiating an automatic reactor scram and automatic
closure of the MSIVs are justified based on the NRC-approved LTR
NEDO-31400A and supporting dose analysis. The supporting dose
analysis also supports the elimination of the MSL drain isolation
function of the MSLRMs on the basis that with the valves open the
source term associated with the analyzed release is directed to the
main condenser the same as it would be via the MSLs themselves.
The methods of analysis and assumptions used to evaluate the
consequences of the applicable impacted safety analysis (i.e. the
CRDA) are consistent with the conservative regulatory requirements
and guidance identified in Section 5.1 above [this is a reference to
``Applicable Regulatory Requirements/Criteria'' in DTE August 24,
2017, license amendment request] and establish estimates of the EAB
[exclusion area boundary], LPZ [low population zone], and MCR [main
control room] doses that comply with these criteria. Hence, there is
reasonable assurance that Fermi 2, modified as proposed by this
submittal, will continue to provide sufficient safety margins to
address unanticipated events and to compensate for uncertainties in
accident progression and analysis assumptions and parameters.
Adding requirements for the MVP and GSE high MSL radiation trips
in the Fermi 2 TS will ensure that appropriate measures and
requirements are in place to maintain the operability of these
functions as such that any release of radioactive material from a
gross fuel failure resulting from a CRDA will be contained in the
main condenser and processed through the offgas system.
The proposed changes do not increase system or component
pressures, temperatures, or flowrates for systems designed to
prevent accidents or mitigate the consequences of an accident.
The analyses performed in accordance with the specified NRC-
approved methods and assumptions demonstrate that the removal of the
trip and isolation functions as described will not cause a
significant reduction in the margin of safety, as the resulting
offsite dose consequences are being maintained within regulatory
limits. The proposed changes do not exceed or alter a design basis
or a safety limit for a parameter to be described or established in
the UFSAR [updated final safety analysis report].
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jon P. Christinidis, DTE Energy, Expert
Attorney--Regulatory, 688 WCB, One Energy Plaza, Detroit, MI 48226-
1279.
NRC Branch Chief: David J. Wrona.
Duke Energy Progress, LLC (Duke Energy), Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1 (HNP), Wake and Chatham Counties,
North Carolina
Duke Energy Progress, LLC, Docket No. 50-261, H.B. Robinson Steam
Electric Plant Unit No. 2 (RNP), Darlington County, South Carolina
Date of amendment request: October 19, 2017. A publicly-available
version is in ADAMS under Accession No. ML17292A040.
Description of amendment request: The proposed amendment request
consists of five changes that would revise the Technical Specifications
(TSs) to support the allowance of Duke Energy to self-perform core
reload design and safety analyses. These changes would (1) add the NRC-
approved COPERNIC Topical Report (TR) to the list of TRs for HNP and
RNP; (2) relocate several TS parameters to the Core Operating Limits
Reports for HNP and RNP; (3) revise the RNP TS Moderator Temperature
Coefficient maximum upper limit; (4) revise the HNP TS definition of
Shutdown Margin consistent with Technical Specifications Task Force
(TSTF) Traveler TSTF-248, Revision 0, ``Revise Shutdown Margin
Definition for Stuck Rod Exception'' (ADAMS Accession No. ML040611010);
and (5) revise the RNP and HNP power distribution limits limiting
condition for operation actions and surveillance requirements to allow
operation of a reactor core designed using the DPC-NE-2011-P
[proprietary], ``Nuclear Design Methodology Report for Core Operating
Limits of Westinghouse Reactors,'' methodology. (A redacted version,
designated as DPC-NE-2011, is publicly-available under ADAMS Accession
No. ML16125A420.)
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
COPERNIC
The proposed change adds a topical report for an NRC-reviewed
and approved fuel performance code to the list of topical reports in
RNP and HNP Technical Specifications (TS), which is administrative
in nature and has no impact on a plant configuration or system
performance relied upon to mitigate the consequences of an accident.
The list of topical reports in the TS used to develop the core
operating limits does not impact either the initiation of an
accident or the mitigation of its consequences.
Relocate TS Parameters to the COLR
The proposed change relocates certain cycle-specific core
operating limits from the RNP and HNP TS to the Core Operating
Limits Report (COLR). The cycle-specific values must be calculated
using the NRC approved methodologies listed in the COLR section of
the TS. Because the parameter limits are determined using the NRC
methodologies, they will continue to be within the limit assumed in
the accident analysis. As a result, neither the probability nor the
consequences of any accident previously evaluated will be affected.
RNP MTC TS Change
The proposed change revises the RNP Technical Specification
maximum upper Moderator Temperature Coefficient (MTC) limit.
Revision of the MTC limit does not affect the performance of any
equipment used to mitigate the consequences of an analyzed accident.
There is no impact on the source term or pathways assumed in
accidents previously assumed. No analysis assumptions are violated
and there are no adverse effects on the factors that contribute to
offsite or onsite dose as the result of an accident.
HNP TSTF-248
The proposed change revises the HNP Technical Specification
definition of Shutdown Margin (SDM) consistent with existing NRC-
approved definition. The proposed revision to the SDM definition
will result in analytical flexibility for determining SDM. Revision
of the SDM definition does not affect the performance of any
equipment used to mitigate the consequences of an analyzed accident.
There is no impact on the source term or pathways assumed in
accidents previously assumed. No analysis assumptions are violated
and there are no adverse effects on the factors that contribute to
offsite or onsite dose as the result of an accident.
DPC-NE-2011-P TS Changes
The proposed change revises the RNP and HNP TS to allow
operation of a reactor core designed using the DPC-NE-2011-P
methodology. The DPC-NE-2011-P methodology has already been approved
by the NRC for use at RNP and HNP. Revision of the TS to align with
the NRC-approved methodology does not affect the performance of any
equipment used to mitigate the consequences of an analyzed accident.
There is no impact on the source term or pathways assumed in
accidents previously assumed.
[[Page 167]]
No analysis assumptions are violated and there are no adverse
effects on the factors that contribute to offsite or onsite dose as
the result of an accident.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
COPERNIC
The proposed change adds a topical report for an NRC-reviewed
and approved fuel performance code to the list of topical reports in
HNP and RNP TS, which is administrative in nature and has no impact
on a plant configuration or on system performance. The proposed
change updates the list of NRC-approved topical reports used to
develop the core operating limits. There is no change to the
parameters within which the plant is normally operated. The
possibility of a new or different kind of accident is not created.
Relocate TS Parameters to the COLR
The proposed change relocates certain cycle-specific core
operating limits from the RNP and HNP TS to the COLR. No new or
different accidents result from utilizing the proposed change. The
changes do not involve a physical alteration of the plant (i.e., no
new or different type of equipment will be installed) or a change in
the methods governing normal plant operation. In addition, the
changes do not impose any new or different requirements or eliminate
any existing requirements. The changes do not alter assumptions made
in the safety analyses. The proposed changes are consistent with the
safety analyses assumptions and current plant operating practice.
RNP MTC TS Change
The proposed change revises the RNP Technical Specification
maximum upper MTC limit. The proposed change does not physically
alter the plant; that is, no new or different type of equipment will
be installed. Therefore the proposed change could also not initiate
an equipment malfunction that would result in a new or different
type of accident from any previously evaluated. This change does not
create new failure modes or mechanisms which are not identifiable
during testing, and no new accident precursors are generated.
HNP TSTF-248
Revising the HNP Technical Specification definition of SDM would
not require revision to any SDM boron calculations. Rather, it would
afford the analytical flexibility for determining SDM for a
particular circumstance. The proposed change does not physically
alter the plant; that is, no new or different type of equipment will
be installed. Therefore the proposed change could also not initiate
an equipment malfunction that would result in a new or different
type of accident from any previously evaluated. This change does not
create new failure modes or mechanisms which are not identifiable
during testing, and no new accident precursors are generated.
DPC-NE-2011-P TS Changes
The proposed change revises the RNP and HNP TS to allow
operation of a reactor core designed using the DPC-NE-2011-P
methodology. The DPC-NE-2011-P methodology has already been approved
by the NRC for use at RNP and HNP. The proposed change does not
physically alter the plant, that is, no new or different type of
equipment will be installed. Therefore the proposed change could
also not initiate an equipment malfunction that would result in a
new or different type of accident from any previously evaluated.
Operating the reactor in accordance with the NRC-approved
methodology will ensure that the core will operate within safe
limits. This change does not create new failure modes or mechanisms
which are not identifiable during testing, and no new accident
precursors are generated.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Margin of safety is related to the confidence in the ability of
the fission product barriers to perform their design functions
during and following an accident. These barriers include the fuel
cladding, the reactor coolant system, and the containment system.
COPERNIC
The proposed change adds a topical report for an NRC-reviewed
and approved fuel performance code to the list of topical reports in
HNP and RNP TS, which is administrative in nature and does not amend
the cycle specific parameters presently required by the TS. The
individual TS continue to require operation of the plant within the
bounds of the limits specified in the COLR. The proposed change to
the list of analytical methods referenced in the COLR does not
impact the margin of safety.
Relocate TS Parameters to the COLR
The proposed change relocates certain cycle-specific core
operating limits from the RNP and HNP TS to the COLR. This change
will have no effect on the margin of safety. The relocated cycle-
specific parameters will continue to be calculated using NRC-
approved methodologies and will provide the same margin of safety as
the values currently located in the TS.
RNP MTC TS Change
The proposed change revises the RNP Technical Specification
maximum upper MTC limit. The MTC limit change does not impact the
reliability of the fission product barriers to function.
Radiological dose to plant operators or to the public will not be
impacted as a result of the proposed change. The current Updated
Final Safety Analysis Report (UFSAR) Chapter 15 analyses of record
remain bounding with the proposed change to the maximum upper MTC
limit. Therefore, all of the applicable acceptance criteria continue
to be met for each of the analyses with the revised maximum upper
MTC limit.
HNP TSTF-248
The proposed revision to the HNP Technical Specification
definition of SDM does not impact the reliability of the fission
product barriers to function. Radiological dose to plant operators
or to the public will not be impacted as a result of the proposed
change. Adequate SDM will continue to be ensured for all operational
conditions.
DPC-NE-2011-P TS Changes
The proposed change revises the RNP and HNP TS to allow
operation of a reactor core designed using the DPC-NE-2011-P
methodology. As a portion of the overall Duke Energy methodology for
cycle reload safety analyses, DPC-NE-2011-P has already been
approved by the NRC for use at RNP and HNP. The proposed change will
continue to ensure that applicable design and safety limits are
satisfied such that the fission product barriers will continue to
perform their design functions. Operation of the reactor in
accordance with the DPC-NE-2011-P methodology will ensure the margin
of safety is not reduced.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Kathryn B. Nolan, Deputy General Counsel,
Duke Energy Corporation, 550 South Tryon Street, Mail Code DEC45A,
Charlotte NC 28202.
NRC Branch Chief: Undine Shoop.
Duke Energy Progress, LLC, Docket No. 50-400, Shearon Harris Nuclear
Power Plant, Unit 1, Wake and Chatham Counties, North Carolina
Date of amendment request: October 10, 2017. A publicly-available
version is in ADAMS under Accession No. ML17283A159.
Description of amendment request: The amendment would revise the
Shearon Harris Nuclear Power Plant (HNP), Unit 1, Technical
Specifications (TSs) to align more closely to improved Standard
Technical Specifications for rod control and to the initial conditions
in the HNP safety analyses. The proposed changes will delete TS action
statement requirements that include a plant shutdown to address rods
that are immovable but trippable. Revisions to surveillance
requirements (SRs) are proposed to clarify actions that are not
necessary if rods are immovable but still trippable.
[[Page 168]]
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Response: No.
The proposed activity will delete action statement 3.1.3.1.c
from the HNP TS and amend action statement 3.1.3.1.d, SR
4.1.1.1.1.a, and SR 4.1.1.2.a. These TS actions address electrical
problems that prevent the Control Rod Drive Mechanism (CRDM) from
moving rods. These conditions do not affect the safety functions of
the control rods or shutdown margin of the unit. Rods will still
insert into the core on an interruption of power to the CRDM, as
occurs in a reactor trip. Also, rod alignment is not impacted,
ensuring no change to reactivity.
The proposed activity is removing actions from the HNP TS for
conditions that do not impact the plant's safety analysis. Rods will
still insert into the core on an interruption of power to the CRDM,
as occurs in a reactor trip. Also, rod alignment is not impacted,
ensuring no change to reactivity or shutdown margin. Since the
conditions of these TS actions do not impact the plant safety
analysis, the plant shutdown directed by them is unnecessary. The
overall probability or consequence of an accident will not be
significantly increased by removing the unnecessary TS actions.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed license amendment create the possibility of
a new or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed activity will delete action statement 3.1.3.1.c
from the HNP TS and amend action statements 3.1.3.1.d, SR
4.1.1.1.1.a, and SR 4.1.1.2.a. These TS actions address electrical
problems that prevent the CRDM from moving rods. These conditions do
not affect the safety functions of the control rods. Rods will still
insert into the core on an interruption of power to the CRDM, as
occurs in a reactor trip. Also, rod alignment is not impacted,
ensuring no change to reactivity or shutdown margin.
The proposed change does not involve installation of new
equipment or modification of existing equipment, so that no new
equipment failure modes are introduced. Also, the proposed change in
TS does not result in a change to the way that the equipment or
facility is operated that would create new accident initiators.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed license amendment involve a significant
reduction in a margin of safety?
Response: No.
The proposed activity will delete action statement 3.1.3.1.c
from the HNP TS and amend action statement 3.1.3.1.d, SR
4.1.1.1.1.a, and SR 4.1.1.2.a. These actions address electrical
problems that prevent the CRDM from moving rods. These conditions do
not affect the safety functions of the control rods. Rods will still
insert into the core on an interruption of power to the CRDM, as
occurs in a reactor trip. Also, rod alignment is not impacted,
ensuring no change to reactivity or shutdown margin.
The TS action statements as amended will continue to address the
two required safety functions of rod control: to shut down the
reactor in the event of a reactor trip, or to maintain proper
alignment to ensure even power distribution. TS action statement
3.1.3.1.a will remain to drive actions if untrippable rods are
identified. TS action statements 3.1.3.1.b and 3.1.3.1.d will remain
to drive actions if misaligned rods are identified. The proposed
changes to HNP TS do not significantly impact either rod safety
function, and separate TS action statements for both functions will
remain in place. Further, the impacted surveillances will continue
to be applicable to conditions impacting either rod safety function.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Kathryn B. Nolan, Deputy General Counsel,
Duke Energy Corporation, 550 South Tryon St., M/C DEC45A, Charlotte, NC
28202.
NRC Branch Chief: Undine Shoop.
Exelon Generation Company, LLC, Docket No. 50-244, R.E. Ginna Nuclear
Power Plant, Wayne County, New York
Date of amendment request: October 31, 2017. A publicly available
version is in ADAMS under Accession No. ML17304A984.
Description of amendment request: The amendment would revise
Technical Specification (TS) Surveillance Requirement 3.8.4.3, ``DC
[Direct Current] Sources--MODES 1, 2, 3, and 4,'' for the R.E. Ginna
Nuclear Power Plant (Ginna). The proposed change would allow the use of
a consistent battery testing technique in order to provide consistent
data for trending battery performance. This proposed change is based on
guidance provided in the Institute of Electrical and Electronics
Engineers (IEEE) Standard 450-2010, ``IEEE Recommended Practice for
Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for
Stationary Applications,'' which is endorsed by NRC Regulatory Guide
1.129, Revision 3, ``Maintenance, Testing, and Replacement of Vented
Lead-Acid Storage Batteries for Nuclear Power Plants.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed change will continue to ensure that the DC system is
tested in a manner that will verify operability. Performance of the
required system surveillances, in conjunction with the applicable
operational and design requirements for the DC system, provide
assurance that the system will be capable of performing the required
design functions for accident mitigation and also that the system
will perform in accordance with the functional requirements for the
system as described in the Updated Final Safety Analysis Report for
Ginna. This change is in accordance with IEEE Standard 450-2010,
``IEEE Recommended Practice for Maintenance, Testing, and
Replacement of Vented Lead-Acid Batteries for Stationary
Applications,'' which has been endorsed by NRC Regulatory Guide
1.129, Revision 3, ``Maintenance, Testing, and Replacement of Vented
Lead-Acid Storage Batteries for Nuclear Power Plants.'' This endures
that the rate of occurrence and consequences of analyzed accidents
will not change.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not create the possibility of a new or
different kind of accident from any previously evaluated. The
proposed surveillance requirement change will continue to ensure
that the DC system and in particular the batteries are tested in a
manner that will verify operability. No physical changes to the
Ginna systems, structures, or components are being implemented.
There are no new or different accident initiators or sequences being
created by the proposed TS change. Therefore, the change does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change does not involve a significant reduction in
the margin of safety.
[[Page 169]]
The proposed DC system surveillance requirement change provides
appropriate and applicable surveillances for the DC system. The
proposed change to surveillance requirements for the DC system will
continue to ensure system operability.
Therefore, this change does not affect any margin of safety for
Ginna.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Tamra Domeyer, Associate General Counsel,
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL
60555.
NRC Branch Chief: James G. Danna.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units Nos. 1 and 2, Berrien County, Michigan
Date of amendment request: November 7, 2017. A publicly-available
version is in ADAMS under Accession No. ML17317A472.
Description of amendment request: The proposed change would allow
for deviation from National Fire Protection Association (NFPA) 805
requirements to allow for currently installed non-plenum listed cables
routed above suspended ceilings and to allow for the use of thin wall
electrical metallic tubing (EMT) and embedded/buried plastic conduit.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The use of EMT and embedded/buried PVC [polyvinyl chloride] does
not create ignition sources and does not impact fire prevention. The
EMT and embedded PVC had been in use since original plant
construction, are allowed by the National Electrical Code and are
not expected to increase the potential for a fire to start.
The prior introduction of non-listed communication/data cables
routed above suspended ceilings does not create ignition sources and
does not impact fire prevention. Cable installation procedures are
utilized to prevent the future installation of new cables that are
noncompliant. Also, the communication/data cables routed above
suspended ceilings do not result in compromising automatic fire
suppression functions, manual fire suppression functions, fire
protection or systems and structures, or post-fire safe shutdown
capability.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes do allow future physical changes to the
facility that deviate from NFPA 805 requirements. However, the
proposed changes do not alter any assumptions made in the safety
analyses, nor do they involve any changes to plant procedures for
ensuring that the plant is operated within analyzed limits. As such,
no new failure modes or mechanisms that could cause a new or
different kind of accident from any previously evaluated are being
introduced.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed changes do not alter the manner in which safety
limits or limiting safety system settings are determined. No changes
to instrument/system actuation setpoints are involved. The safety
analysis acceptance criteria are not affected by this change and the
proposed changes will not permit plant operation in a configuration
outside the design basis.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Robert B. Haemer, Senior Nuclear Counsel,
One Cook Place, Bridgman, MI 49106.
NRC Branch Chief: David J. Wrona.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant (CNP), Units Nos. 1 and 2, Berrien County,
Michigan
Date of amendment request: November 7, 2017. A publicly-available
version is in ADAMS under Package Accession No. ML17317A454.
Description of amendment request: The proposed change would revise
the CNP Emergency Plan to relocate the Technical Support Center (TSC)
within the CNP protected area.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change to the CNP emergency plan to relocate the
TSC does not impact the physical function of plant structures,
systems, or components (SSC) or the manner in which SSCs perform
their design function. The proposed change neither adversely affects
accident initiators or precursors, nor alters design assumptions.
The proposed change does not alter or prevent the ability of SSCs to
perform their intended function to mitigate the consequences of an
initiating event within assumed acceptance limits. No operating
procedures or administrative controls that function to prevent or
mitigate accidents are affected by the proposed changes.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change does not impact the accident analysis. The
proposed change does not involve a physical alteration of the plant
(i.e., no new or different type of equipment will be installed or
removed) or a change in the method of plant operation. The proposed
change will not introduce failure modes that could result in a new
accident, and the change does not alter assumptions made in the
safety analysis. The proposed change to the location of the TSC is
not an initiator of any accidents.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Margin of safety is associated with confidence in the ability of
the fission product barriers (i.e., fuel cladding, reactor coolant
system pressure boundary, and containment structure) to limit the
level of radiation dose to the public. The proposed change does not
impact operation of the plant or its response to transients or
accidents. The change does not affect the Technical Specifications
or the operating license other than to amend the license to approve
the change. The proposed change does not involve a change in the
method of plant operation, and no accident analyses will be affected
by the proposed changes.
Additionally, the proposed change will not relax any criteria
used to establish safety limits and will not relax any safety system
settings. The safety analysis acceptance criteria are not affected
by these changes. The proposed change will not result in plant
operation in a configuration outside the design basis. The proposed
change does not adversely affect systems that respond to safely shut
down the plant and to maintain
[[Page 170]]
the plant in a safe shutdown condition. The emergency plan will
continue to activate an emergency response commensurate with the
extend of degradation of plant safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Robert B. Haemer, Senior Nuclear Counsel,
One Cook Place, Bridgman, MI 49106.
NRC Branch Chief: David J. Wrona.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026,
Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County,
Georgia
Date of amendment request: October 6, 2017. A publicly-available
version is in ADAMS under Accession No. ML17279B017.
Description of amendment request: The requested amendment proposes
changes to the licensing basis documents to change the methodology and
acceptance criteria for the in-containment refueling water storage tank
(IRWST) heatup preoperational test described in the Updated Final
Safety Analysis Report (UFSAR) Subsection 14.2.9.1.3, item h, and the
passive residual heat removal (PRHR) heat exchanger preoperational test
described in UFSAR Subsection 14.2.9.1.3, item g. These changes involve
material which is specifically referenced in Section 2.D.(2) of the
combined licenses for VEGP, Units 3 and 4.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This activity changes the acceptance criteria for the IRWST
heatup preoperational test and provides allowance to perform the
preoperational test during both PRHR heat exchanger natural
circulation and forced flow, instead of only during natural
circulation. In addition, the acceptance criteria are changed for
the PRHR heat exchanger forced flow system operability and
preoperational tests.
No structure, system, or component (SSC) or function is changed
by this proposed activity. There is no change to the application of
Regulatory Guide 1.68, nor is there a change to the design of the
PRHR heat exchanger or the IRWST. The initial test program continues
to confirm the heat transfer capability of the PRHR heat exchanger
and that the IRWST heatup is consistent with the PRHR heat exchanger
heat transfer modeling in the UFSAR Chapter 15 safety analysis.
The proposed amendment does not affect the prevention or
mitigation of abnormal events; e.g., accidents, anticipated
operation occurrences, earthquakes, floods, turbine missiles, and
fires or their safety or design analyses. This change does not
involve containment of radioactive isotopes or have any adverse
effect on a fission product barrier. There is no impact on
previously evaluated accidents.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed changes do not involve a new failure mechanism or
malfunction, that affects an SSC accident initiator, or interface
with any SSC accident initiator or initiating sequence of events
considered in the design and licensing bases. There is no adverse
effect on radioisotope barriers or the release of radioactive
materials. The proposed amendment does not adversely affect any
accident, including the possibility of creating a new or different
kind of accident from any accident previously evaluated. Therefore,
the proposed amendment does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
This activity changes the acceptance criteria for the IRWST
heatup preoperational test and gives allowance to perform the
preoperational test during both PRHR heat exchanger natural
circulation and forced flow, instead of only during natural
circulation. In addition, the acceptance criteria are changed for
the PRHR heat exchanger forced flow system operability and
preoperational tests.
No SSC or function is changed within this activity. There is no
change to the application of Regulatory Guide 1.68, nor is there a
change to how the PRHR heat exchanger or the IRWST are designed. The
initial test program continues to confirm the heat transfer
capability of the PRHR heat exchanger. The initial test program will
confirm the IRWST heatup is consistent with the current PRHR heat
exchanger heat transfer modeling in the UFSAR Chapter 15 safety
analysis.
The proposed changes would not affect any safety-related design
code, function, design analysis, safety analysis input or result, or
existing design/safety margin. No safety analysis or design basis
acceptance limit/criterion is challenged or exceeded by the
requested changes.
Therefore, the requested amendment does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP,
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Jennifer Dixon-Herrity.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026,
Vogtle Electric Generating Plant, Units 3 and 4, Burke County, Georgia
Date of amendment request: November 16, 2017. A publicly-available
version is in ADAMS under Accession No. ML17325A562.
Description of amendment request: The amendments propose changes to
Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) in
Combined License (COL) Appendix C, with corresponding changes to the
associated plant-specific Tier 1 information to simplify and
consolidate a number of ITAAC to improve efficiency of the ITAAC
completion and closure process. Pursuant to the provisions of 10 CFR
52.63(b)(1), an exemption from elements of the design as certified in
the 10 CFR part 52, Appendix D, design certification rule is also
requested for the plant-specific Design Control Document Tier 1
material departures.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed non-technical change to COL Appendix C will
consolidate ITAAC in order to improve and create a more efficient
process for the ITAAC Closure Notification submittals. No structure,
system, or component (SSC) design or function is affected. No design
or safety analysis is affected. The proposed changes do not affect
any accident initiating event or component failure, thus the
probabilities of the accidents previously evaluated are not
affected. No function used to mitigate a radioactive material
release and no radioactive material release source term is involved,
thus the
[[Page 171]]
radiological releases in the accident analyses are not affected.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change to COL Appendix C does not affect the design
or function of any SSC, but will consolidate ITAAC in order to
improve efficiency of the ITAAC completion and closure process. The
proposed changes would not introduce a new failure mode, fault or
sequence of events that could result in a radioactive material
release.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change to COL Appendix C to consolidate ITAAC in
order to improve efficiency of the ITAAC completion and closure
process is considered non-technical and would not affect any design
parameter, function or analysis. There would be no change to an
existing design basis, design function, regulatory criterion, or
analysis. No safety analysis or design basis acceptance limit/
criterion is involved.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M. Stanford Blanton, Balch & Bingham LLP,
1710 Sixth Avenue North, Birmingham, AL 35203-2015.
NRC Branch Chief: Jennifer Dixon-Herrity.
Tennessee Valley Authority, Docket No. 50-391, Watts Bar Nuclear Plant,
Unit 2, Rhea County, Tennessee
Date of amendment request: October 11, 2017. A publicly-available
version is in ADAMS under Accession No. ML17284A452.
Description of amendment request: The amendment would revise
Technical Specification (TS) 3.3.1, Table 3.3.1-1, ``Reactor Trip
System (RTS) Instrumentation,'' to increase the values for the nominal
trip setpoint and the allowable value for Function 14.a, ``Turbine
Trip--Low Fluid Oil Pressure.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change reflects a design change to the turbine
control system that results in the use of an increased control oil
pressure system, necessitating a change to the value at which a low
fluid oil pressure initiates a reactor trip on turbine trip. The low
fluid oil pressure is an input to the reactor trip instrumentation
in response to a turbine trip event. The value at which the low
fluid oil initiates a reactor trip is not an accident initiator. A
change in the nominal control oil pressure does not introduce any
mechanisms that would increase the probability of an accident
previously analyzed. The reactor trip on turbine trip function is
initiated by the same protective signal as used for the existing
auto stop low fluid oil system trip signal. There is no change in
form or function of this signal and the probability or consequences
of previously analyzed accidents are not impacted.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the [proposed] change create the possibility of a new or
different kind of accident from any previously evaluated?
Response: No.
The EHC [electrohydraulic control] fluid oil pressure rapidly
decreases in response to a turbine trip signal. The value at which
the low fluid oil pressure switches initiates a reactor trip is not
an accident initiator. The proposed TS change reflects the higher
pressure that will be sensed after the pressure switches are
relocated from the auto stop low fluid oil system to the EHC high
pressure header. Failure of the new switches would not result in a
different outcome than is considered in the current design basis.
Further, the change does not alter assumptions made in the safety
analysis but ensures that the instruments perform as assumed in the
accident analysis.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the [proposed] change involve a significant reduction in
a margin of safety?
Response: No.
The change involves a parameter that initiates an anticipatory
reactor trip following a turbine trip. The safety analyses do not
credit this anticipatory trip for reactor core protection. The
original pressure switch configuration and the new pressure switch
configuration both generate the same reactor trip signal. The
difference is that the initiation of the trip will now be adjusted
to a different system of higher pressure. This system function of
sensing and transmitting a reactor trip signal on turbine trip
remains the same. There is no impact to safety analysis acceptance
criteria as described in the plant licensing basis because no change
is made to the accident analysis assumptions.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Sherry A. Quirk, Executive Vice President
and General Counsel, Tennessee Valley Authority, 400 West Summit Hill
Drive, 6A West Tower, Knoxville, TN 37902.
NRC Branch Chief: Undine Shoop.
III. Notice of Issuance of Amendments to Facility Operating Licenses
and Combined Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR chapter I, which are set
forth in the license amendment.
A notice of consideration of issuance of amendment to facility
operating license or combined license, as applicable, proposed no
significant hazards consideration determination, and opportunity for a
hearing in connection with these actions, was published in the Federal
Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items can be accessed as described in
[[Page 172]]
the ``Obtaining Information and Submitting Comments'' section of this
document.
Entergy Nuclear Operations, Inc., Docket Nos. 50-003, 50-247, and 50-
286, Indian Point Nuclear Generating Unit Nos. 1, 2, and 3, Westchester
County, New York
Date of amendment request: April 28, 2017, as supplemented by
letters dated August 9, 2017; September 28, 2017; and October 26, 2017.
Brief description of amendments: The amendments revised the Cyber
Security Plan Milestone 8 full implementation date by extending the
full implementation date from December 31, 2017, to December 31, 2018.
Date of issuance: December 8, 2017.
Effective date: As of the date of issuance, and shall be
implemented by December 31, 2017.
Amendment Nos.: 60 (Unit No. 1), 286 (Unit No. 2), and 263 (Unit
No. 3). A publicly-available version is in ADAMS under Accession No.
ML17315A000; documents related to these amendments are listed in the
Safety Evaluation enclosed with the amendments. Provisional Operating
License No. DPR-5 and Facility Operating License Nos. DPR-26 and DPR-
64: The amendments revised the Provisional Operating License for Unit
No. 1 and the Facility Operating Licenses for Unit Nos. 2 and 3.
Date of initial notice in Federal Register: July 18, 2017 (82 FR
32880).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 8, 2017.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-255, Palisades Nuclear
Plant, Van Buren County, Michigan
Date of amendment request: March 30, 2017, as supplemented by
letter dated October 17, 2017.
Brief description of amendment: This amendment revised the Cyber
Security Plan (CSP) implementation schedule Milestone 8 date and
paragraph 2.E in the renewed facility operating license from December
15, 2017, to March 31, 2019. Milestone 8 of the CSP implementation
schedule concerns the full implementation of the CSP.
Date of issuance: December 15, 2017.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 264. A publicly-available version is in ADAMS under
Accession No. ML17328B033; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-20: Amendment revised
the Renewed Facility Operating License.
Date of initial notice in Federal Register: May 23, 2017 (82 FR
23623). The supplemental letter dated October 17, 2017, provided
additional information that expanded the scope of the application as
originally noticed and changed the NRC staff's original proposed no
significant hazards consideration (NSHC) determination as published in
the Federal Register. Accordingly, the NRC published a second proposed
NSHC determination in the Federal Register on November 7, 2017 (82 FR
51650). This notice superseded the original notice in its entirety. It
also provided an opportunity to request a hearing by January 8, 2018,
but indicated that if the Commission makes a final NSHC determination,
any such hearing would take place after issuance of the amendment.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 15, 2017.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station (Pilgrim), Plymouth County, Massachusetts
Date of amendment request: March 30, 2017.
Brief description of amendment: The amendment revised Pilgrim's
renewed facility operating license for the Cyber Security Plan (CSP)
Milestone 8 full implementation completion date, as set forth in the
CSP implementation schedule, and revised the physical protection
license condition. The amendment revised the CSP Milestone 8 completion
date from December 15, 2017, to December 31, 2020.
Date of issuance: December 15, 2017.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 247. A publicly-available version is in ADAMS under
Accession No. ML17290A487; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-35: The amendment
revised the renewed facility operating license.
Date of initial notice in Federal Register: May 23, 2017 (82 FR
23624).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 15, 2017.
No significant hazards consideration comments received: No.
National Institute of Standard and Technology (NIST), Docket No. 50-
184, National Bureau of Standards Test Reactor (NBSR), Montgomery
County, Maryland
Date of amendment request: March 2, 2017, as supplemented by
letters dated March 29, 2017; May 25, 2017; November 17, 2017; November
20, 2017; December 1, 2017; December 11, 2017; and December 14, 2017.
Brief description of amendment: The amendment revised NIST NBSR's
Facility Operating License TR-5 to allow receipt of calibration and
testing sources, and revised technical specifications pertaining to the
NIST reactor low power startup testing and organizational reporting
requirements.
Date of issuance: December 15, 2017.
Effective date: As of the date of issuance.
Amendment No.: 11. A publicly-available version is in ADAMS under
Accession No. ML17292A062; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. TR-5: Amendment revised the Renewed
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: September 12, 2017 (82
FR 42844). The supplemental letters dated November 17, 2017; November
20, 2017; December 1, 2017; December 11, 2017; and December 14, 2017
(which withdrew parts of the application), provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the NRC
staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 15, 2017.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit 1 (FCS), Washington County, Nebraska
Date of amendment request: December 16, 2016, as supplemented by
letter dated May 15, 2017.
Brief description of amendment: The amendment revised the FCS
Emergency Plan and Emergency Action Level (EAL) scheme for the
permanently defueled condition. The proposed permanently defueled
Emergency Plan and EAL scheme are commensurate with the
[[Page 173]]
significantly reduced spectrum of credible accidents that can occur in
the permanently defueled condition and are necessary to properly
reflect the conditions of the facility while continuing to preserve the
effectiveness of the emergency plan.
Date of issuance: December 12, 2017.
Effective date: The amendment is effective April 7, 2018, and shall
be implemented within 90 days of the effective date.
Amendment No.: 295. A publicly-available version is in ADAMS under
Accession No. ML17276B286; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. DPR-40: The amendment
revised the Emergency Plan and EAL scheme.
Date of initial notice in Federal Register: March 28, 2017 (82 FR
15383). The supplemental letter dated May 15, 2017, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the original
proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 12, 2017.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: March 27, 2017.
Brief description of amendment: The licensee requested to adopt
NRC-approved Technical Specifications Task Force (TSTF) Improved
Standard Technical Specifications Change Traveler TSTF-535, Revision 0,
``Revise Shutdown Margin Definition to Address Advanced Fuel Designs''
(ADAMS Accession No. ML112200436), dated August 8, 2011. The definition
of shutdown margin in the Hope Creek Generating Station Technical
Specifications is revised to require calculation of shutdown margin at
the reactor moderator temperature corresponding to the most reactive
state throughout the operating cycle, which is 68 degrees Fahrenheit or
higher.
Date of issuance: December 13, 2017.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 208. A publicly-available version is in ADAMS under
Accession No. ML17317A605; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-57: Amendment revised
the Renewed Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: May 9, 2017 (82 FR
21560).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 13, 2017.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: March 27, 2017, as supplemented by
letters dated April 28, 2017, and September 5, 2017.
Brief description of amendment: The amendment changed the Hope
Creek Generating Station Technical Specifications (TSs) to relocate the
reactor coolant system pressure-temperature (P-T) limit curves from the
TSs to a new licensee-controlled document called the Pressure and
Temperature Limits Report. The amendment also revised the 32 effective
full power years P-T limit curves and approved P-T limit curves
applicable through the license renewal term. The revisions to the
curves were required due to the results of a recently pulled and tested
reactor pressure vessel surveillance capsule.
Date of issuance: December 14, 2017.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 209. A publicly-available version is in ADAMS under
Accession No. ML17324A840; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Renewed Facility Operating License No. NPF-57: Amendment revised
the Renewed Facility Operating License and TSs.
Date of initial notice in Federal Register: May 23, 2017 (82 FR
23628). The supplemental letter dated September 5, 2017, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the NRC staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 14, 2017.
No significant hazards consideration comments received: No.
Southern California Edison Company, et al., Docket Nos. 50-206, 50-361,
and 50-362, San Onofre Nuclear Generating Station, Units 1, 2, and 3,
San Diego County, California
Date of amendment request: December 15, 2016, as supplemented by
letter dated May 5, 2017.
Brief description of amendments: The amendments replaced the San
Onofre Nuclear Generating Station, Units 1, 2, and 3 (SONGS)
Permanently Defueled Emergency Plan and associated Emergency Action
Level (EAL) Bases Manual (hereafter referred to as the EAL scheme) with
an Independent Spent Fuel Storage Installation (ISFSI) Only Emergency
Plan (IOEP) and associated EAL scheme. The NRC staff determined that
the proposed SONGS IOEP and associated EAL changes continue to meet the
standards in 10 CFR 50.47, ``Emergency plans,'' and the requirements in
Appendix E, ``Emergency Planning and Preparedness for Production and
Utilization Facilities,'' of 10 CFR part 50, as exempted. As such, the
SONGS IOEP and associated EAL changes provide reasonable assurance that
adequate protective measures can and will be taken in the event of a
radiological emergency. These changes more fully reflect the status of
the facility, as well as the reduced scope of potential radiological
accidents once all spent fuel has been moved to dry cask storage within
the onsite ISFSI, an activity which is currently scheduled for
completion in 2019.
Date of issuance: November 30, 2017.
Effective date: As of the date Southern California Edison submits a
written notification to the NRC that all spent nuclear fuel assemblies
have been transferred out of the SONGS spent fuel pools and placed in
storage within the onsite ISFSI, and shall be implemented within 60
days.
Amendment Nos.: 168 (Unit 1), 236 (Unit 2), and 229 (Unit 3). A
publicly-available version is in ADAMS under Accession No. ML17310B482;
documents related to these amendments are listed in the Safety
Evaluation enclosed with the amendments.
Facility Operating License Nos. DPR-13, NPF-10, and NPF-15: The
amendments revised the Facility Operating Licenses.
Date of initial notice in Federal Register: February 14, 2017 (82
FR 10601).
The Commission's related evaluation of the amendments is contained
in a
[[Page 174]]
Safety Evaluation dated November 30, 2017.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026,
Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County,
Georgia
Date of amendment request: May 10, 2017, and supplemented by letter
dated September 20, 2017.
Description of amendments: The amendments consisted of changes to
the VEGP, Units 3 and 4, Updated Final Safety Analysis Report in the
form of departures from the incorporated plant-specific Design Control
Document Tier 2* and Tier 2 information (text, tables, and figures).
Specifically, the amendments consisted of changes related to revising
the design reinforcement in the roof of the auxiliary building and the
design of the girders supporting the roof.
Date of issuance: December 5, 2017.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment Nos.: 101 (Unit 3) and 100 (Unit 4). A publicly-available
version is in ADAMS under Package Accession No. ML17311B236; documents
related to these amendments are listed in the Safety Evaluation
enclosed with the amendments.
Facility Combined Licenses No. NPF-91 and NPF-92: Amendments
revised the Facility Combined Licenses.
Date of initial notice in Federal Register: June 6, 2017 (82 FR
26137). The supplemental letter dated September 20, 2017, provided
additional information that clarified the application, did not expand
the scope of the application request as originally noticed, and did not
change the NRC staff's original proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 5, 2017.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026,
Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County,
Georgia
Date of amendment request: June 23, 2017.
Description of amendments: The amendments consisted of changes to
the VEGP, Units 3 and 4, Updated Final Safety Analysis Report (UFSAR)
in the form of departures from the plant-specific Design Control
Document Tier 2 information and involves changes to the VEGP, Units 3
and 4, Combined License Appendix A, Technical Specifications (TSs).
Specifically, the proposed changes revise plant-specific Tier 2
information to add the time delay assumed in the safety analysis for
the reactor trip on a safeguards actuation (``S'') signal to UFSAR
Table 15.0-4a. This is also reflected in the proposed revision to TS
3.3.4, ``Reactor Trip System (RTS) Engineered Safety Feature Actuation
System (ESFAS) Instrumentation,'' to add a surveillance requirement to
verify the RTS response time for this ``S'' signal. The request also
includes proposed changes to TS 3.3.7, ``RTS Trip Actuation Devices,''
to clarify that the requirements for reactor trip breaker (RTB)
undervoltage and shunt trip mechanisms apply only to in-service RTBs.
In addition, the request includes proposed changes to TS 3.3.9, ``ESFAS
Manual Initiation,'' to correct the nomenclature for the Chemical and
Volume Control System, which is inadvertently stated as the Chemical
Volume and Control System.
Date of issuance: December 8, 2017.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment Nos.: 102 (Unit 3) and 101 (Unit 4). A publicly-available
version is in ADAMS under Accession No. ML17296A236; documents related
to these amendments are listed in the Safety Evaluation enclosed with
the amendments.
Facility Combined Licenses No. NPF-91 and NPF-92: Amendments
revised the Facility Combined Licenses.
Date of initial notice in Federal Register: August 15, 2017 (82 FR
38714).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 8, 2017.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026,
Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County,
Georgia
Date of amendment request: October 20, 2016.
Description of amendments: The amendments authorized changes to the
Tier 2* information in the VEGP, Units 3 and 4, Updated Final Safety
Analysis Report (which includes the plant-specific design control
document information) to clarify the demonstration of the quality and
strength of a specific set of couplers welded to carbon steel embedment
plates, already installed and embedded in concrete through visual
examination and static tension testing, in lieu of the nondestructive
examination requirements of American Institute of Steel Construction
(AISC) N690.
Date of issuance: September 5, 2017.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment Nos.: 86 (Unit 3) and 85 (Unit 4). A publicly-available
version is in ADAMS under Package Accession No. ML17178A197; documents
related to these amendments are listed in the Safety Evaluation
enclosed with the amendments.
Facility Combined Licenses Nos. NPF-91 and NPF-92: Amendments
revised the Facility Combined Licenses.
Date of initial notice in Federal Register: March 14, 2017 (82 FR
13662).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated September 5, 2017.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: March 31, 2017.
Brief description of amendment: The amendment revised Technical
Specification (TS) 5.7.2.14, ``Ventilation Filter Testing Program
(VFTP),'' to correct an administrative error introduced by Amendment
No. 92, issued June 19, 2013. Specifically, Amendment 92 deleted TS
3.9.8, ``Reactor Building Purge Air Cleanup Units,'' but did not delete
associated references to the reactor building purge filters from TS
5.7.2.14.
Date of issuance: December 7, 2017.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 117. A publicly-available version is in ADAMS under
Accession No. ML17311A786; documents related to this amendment are
listed in the Safety Evaluation enclosed with the amendment.
Facility Operating License No. NPF-90: Amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: July 5, 2017 (82 FR
31103).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 7, 2017.
No significant hazards consideration comments received: No.
[[Page 175]]
Dated at Rockville, Maryland, this 21st day of December 2017.
For the Nuclear Regulatory Commission.
Kathryn M. Brock,
Acting Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 2017-27930 Filed 12-29-17; 8:45 am]
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